A diagnostic system for Thomson scattering of the central, edge and divertor plasma regions of a tokamak with reactor technologies is discussed. The rationale and choice of technical solutions are given, the composition of the Thomson scattering diagnostic complex is discussed, as well as an estimate of the accuracy of measuring the electron temperature and plasma density in the central edge and divertor regions of the TRT tokamak. Particular attention is paid to ensuring the functionality of the proposed diagnostics in the reactor mode of the tokamak operation and the results of testing diagnostic equipment in experiments on the Globus-M2 tokamak.

Thomson scattering of the core edge and divertor plasma regions of a tokamak with reactor technologies is discussed. The rationale and choice of technical solutions are given, the composition of the Thomson scattering diagnostic complex is discussed, as well as an estimate of the accuracy of measuring both electron temperature and density. Particular attention is paid to ensuring the functionality of the proposed diagnostics in the reactor mode of the tokamak operation and the results of testing diagnostic equipment in the experiments on Globus-M2 tokamak.

The main results of numerical simulation of transient electromagnetic processes in the modules of the first and second rows of TRT first wall panels are presented. The study was carried out using the TYPHOON software designed for numerical simulation of quasi-stationary eddy currents in conducting shells of complex shape randomly located in space, taking into account their multicoupling and branching. The basic design of the first wall panel and its five modifications are considered. It is determined that the electromagnetic forces and moments for the basic design reach extreme values, and the edge elements of the panel are the most loaded. Options with inserting electrical insulation at the places of contact of the plasma facing elements and the base of the panel, as well as incorporating a shunt between adjacent panels, make it possible to achieve the strongest reduction in loads. These options are accepted as basic for further modification of the panel design and conduction of additional EM analysis.

The concept of next-generation spherical tokamak is being considered: the Globus-3 project, which, in its characteristics, is compatible with the infrastructure existing at the Ioffe Institute, but differs from the currently operating Globus-M2 tokamak in the stronger toroidal magnetic field (1.5–3.0 T) and increased duration of plasma discharge. The parametric analysis data are presented that determined the preliminary selection of the facility parameters. Three options for the electromagnetic system were considered: with the warm copper coils, with the pre-cooled copper coils and with the coils made of high-temperature superconductors. For the first option, the concept for designing the electromagnetic system and vacuum vessel of the facility has been developed. The basic shot scenario with duration of up to 3 s at the field of 1.5 T and plasma current of 0.8 MA is presented.

A neutron collimator is developed to attenuate the neutron flux and reduce the residual induced activity in the interportal space of the diagnostic system of neutral particle analyzers of the ITER tokamak reactor. The collimator is installed in the port plug of the ITER vacuum vessel in front of an inlet to the vacuum pipeline of the diagnostic system. The collimator design has a cellular structure with 80% transparency for the neutral atom beam that goes out of the plasma and is recorded by the analyzers. However, because of increased scattering of neutrons in the collimator, their flux in the interportal room in the service zone of diagnostic systems of equatorial port no. 11 is significantly reduced and allows the equivalent dose rate in this zone to be decreased by several times. Thermal analysis showed that, during the reactor operation in modes with generation of the maximum power of 500 MW, the plasma radiation will cause the heating of the collimating grid to a temperature not exceeding 250°C, which makes it possible to select the stainless steel (316L(N)-ITER grade) as a material for manufacturing the collimator. In this case the cyclic strength of the collimator meets the ITER requirements, and it can be used without replacement during the entire deuterium-tritium experiment of the tokamak reactor.

НЕЙТРОННЫЙ КОЛЛИМАТОР ДЛЯ ДИАГНОСТИЧЕСКОЙ СИСТЕМЫ АТОМНЫХ АНАЛИЗАТОРОВ ТОКАМАКА-РЕАКТОРА ИТЭРА.С.Наволоцкий 1 , В

— The design of a diamond neutral-particle spectrometer (DNPS) is presented. The DNPS is a part of the neutral particle analyzer (NPA) at the International Thermonuclear Experimental Reactor (ITER). The operating modes of the DNPS in the NPA complex, the advantages of the DNPS, and its limitations are described. The DNPS system includes two diamond detectors that differ in sensitivity by a factor of 9. The energy losses in detector contacts are numerically estimated. The test results for the diamond detector are presented.

The control of the deuterium–tritium (DT) fuel isotopic ratio has to ensure the best performance of the ITER thermonuclear fusion reactor. The diagnostic system described in this paper allows the measurement of this ratio analyzing the hydrogen isotope fluxes (performing neutral particle analysis (NPA)). The development and supply of the NPA diagnostics for ITER was delegated to the Russian Federation. The diagnostics is being developed at the Ioffe Institute. The system consists of two analyzers, viz., LENPA (Low Energy Neutral Particle Analyzer) with 10–200 keV energy range and HENPA (High Energy Neutral Particle Analyzer) with 0.1–4.0MeV energy range. Simultaneous operation of both analyzers in different energy ranges enables researchers to measure the DT fuel ratio both in the central burning plasma (thermonuclear burn zone) and at the edge as well. When developing the diagnostic complex, it was necessary to account for the impact of several factors: high levels of neutron and gamma radiation, the direct vacuum connection to the ITER vessel, implying high tritium containment, strict requirements on reliability of all units and mechanisms, and the limited space available for accommodation of the diagnostic hardware at the ITER tokamak. The paper describes the design of the diagnostic complex and the engineering solutions that make it possible to conduct measurements under tokamak reactor conditions. The proposed engineering solutions provide a safe—with respect to thermal and mechanical loads—common vacuum channel for hydrogen isotope atoms to pass to the analyzers; ensure efficient shielding of the analyzers from the ITER stray magnetic field (up to 1 kG); provide the remote control of the NPA diagnostic complex, in particular, connection/disconnection of the NPA vacuum beamline from the ITER vessel; meet the ITER radiation safety requirements; and ensure measurements of the fuel isotopic ratio under high levels of neutron and gamma radiation.

The level of knowledge accumulated to date in the physics and technologies of controlled thermonuclear fusion (CTF) makes it possible to begin designing fusion-fission hybrid systems that would involve a fusion neutron source (FNS) and which would admit employment for the production of fissile materials and for the transmutation of spent nuclear fuel. Modern Russian strategies for CTF development plan the construction to 2023 of tokamak-based demonstration hybrid FNS for implementing steady-state plasma burning, testing hybrid blankets, and evolving nuclear technologies. Work on designing the DEMO-FNS facility is still in its infancy. The Efremov Institute began designing its magnet system and vacuum chamber, while the Kurchatov Institute developed plasma-physics design aspects and determined basic parameters of the facility. The major radius of the plasma in the DEMO-FNS facility is R = 2.75 m, while its minor radius is a = 1 m; the plasma elongation is k (95) = 2. The fusion power is P (FUS) = 40 MW. The toroidal magnetic field on the plasma-filament axis is B (t0) = 5 T. The plasma current is I (p) = 5 MA. The application of superconductors in the magnet system permits drastically reducing the power consumed by its magnets but requires arranging a thick radiation shield between the plasma and magnet system. The central solenoid, toroidal-field coils, and poloidal-field coils are manufactured from, respectively, Nb3Sn, NbTi and Nb3Sn, and NbTi. The vacuum chamber is a double-wall vessel. The space between the walls manufactured from 316L austenitic steel is filled with an iron-water radiation shield (70% of stainless steel and 30% of water).

ОСОБЕННОСТИ КОНСТРУКЦИИ СИСТЕМЫ ДИАГНОСТИКИ ПО ПОТОКАМ АТОМОВ ИЗОТОПОВ ВОДОРОДА ДЛЯ МЕЖДУНАРОДНОГО ТЕРМОЯДЕРНОГО РЕАКТОРА ИТЭРС.Я.Петров 1 , В